NDA Report no. NDA/RWMD/068

Geological Disposal

Generic specification for wastepackages containing low heatgenerating waste

August 2012

NDA Report no. NDA/RWMD/068

Geological Disposal

Generic specification for wastepackages containing low heatgenerating waste

August 2012

ii

Conditions of Publication This report is made available under the NDA Transparency Policy. In line with this policy, the NDA is seeking to make information on its activities readily available, and to enable interested parties to have access to and influence on its future programmes. The report may be freely used for non-commercial purposes. However, all commercial uses, including copying and re-publication, require permission from the NDA. All copyright, database rights and other intellectual property rights reside with the NDA. Applications for permission to use the report commercially should be made to the NDA Information Manager.

Although great care has been taken to ensure the accuracy and completeness of the information contained in this publication, the NDA can not assume any responsibility for consequences that may arise from its use by other parties.

© Nuclear Decommissioning Authority 2012. All rights reserved.

ISBN 978-1-84029-452-1

Bibliography If you would like to see other reports available from NDA, a complete listing can be viewed at our website www.nda.gov.uk, or please write to the Library at the address below.

Feedback Readers are invited to provide feedback to the NDA on the contents, clarity and presentation of this report and on the means of improving the range of NDA reports published. Feedback should be addressed to:

Dr Elizabeth Atherton Head of Stakeholder Engagement and Communications, Nuclear Decommissioning Authority (Radioactive Waste Management Directorate), Building 587 Curie Avenue, Harwell Oxford, Didcot, OX11 0RH, UK

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Executive Summary

We, the Nuclear Decommissioning Authority (NDA), have been charged with implementing Government policy for the long-term management of higher activity radioactive waste, as defined by the Managing Radioactive Waste Safely (MRWS) White Paper, by planning, building and operating a geological disposal facility (GDF) in the UK.

A GDF is an engineered facility for the disposal of radioactive waste. It will be located at a depth of between 200m and 1,000m below ground, in a geology that provides long-term isolation of the wastes from the human environment.

The wastes destined for disposal in a GDF comprise those not considered suitable for near-surface disposal (i.e. high level waste, intermediate level waste (ILW) and some types of low level waste). There are also other nuclear materials that have not been declared as wastes by the Government (because they are still considered to be of potential use), but which might be the subject of geological disposal in the future, namely spent nuclear fuel, separated plutonium and uranium.

A key aspect of the MRWS White Paper is that it envisages that the Radioactive Waste Management Directorate (RWMD) of the NDA will evolve into the organisation responsible for the delivery of a GDF.

As implementer and future operator of a GDF, and therefore as the ultimate receiver of waste for disposal, RWMD will be responsible for the production of waste acceptance criteria (WAC) for the facility. Whilst plans for the construction of a GDF remain at an early stage, the information necessary to define WAC is not available. In the meantime, and as a precursor to the final WAC, we produce generic specifications for packaged waste, the primary purpose of which are to provide a baseline against which the suitability of plans to package waste for disposal can be judged. By providing such a baseline we assist the holders of radioactive waste in the development and implementation of such plans by providing confidence that the resulting waste packages would be compatible with the anticipated needs for transport to and disposal in a GDF.

The purpose of this Generic Specification for waste packages containing low heat generating waste is to define generic requirements for waste packages containing such waste that would be suitable for geological disposal in a manner defined for wastes with radiological properties typical of ILW. The packaging requirements defined by this Generic Specification are derived from the high-level requirements defined by the Generic Waste Package Specification, as applied to waste packages containing such wastes.

This Generic Specification also acts as the basis for the definition of the Waste Package Specifications which define the requirements for the waste packages containing low heat generating waste that would result from the use of standardised designs of waste container.

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List of Contents

Executive Summary iii

1 Introduction 1

2 Background 3

2.1 The geological disposal of radioactive waste 3

2.2 The Disposal System Specification 3

2.3 The Disposal System Safety Case 4

2.4 Specifications for packaged waste 5

2.5 Assessing the disposability of waste packages 5

3 The wastes covered by this Generic Specification 8

4 The development of this Generic Specification 10

5 Basis for the definition of the packaging requirements 12

5.1 The transport of waste packages to a GDF 14

5.2 The disposal of waste packages in a GDF 16

6 Requirements for waste packages containing low heat generating waste 17

6.1 Requirements for waste containers 18

6.2 Requirements for wasteforms 23

6.3 Requirements for waste packages 24

6.4 Requirements for the manufacture and storage of waste packages 34

7 Summary 36

References 37

Appendix A Glossary of terms used in this document 40

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Abbreviations and acronyms used in this document

ALARP as low as reasonably practicable

BSL Basic Safety Level

BSO Basic Safety Objective

DSS Disposal System Specification

DSSC Disposal System Safety Case

DSFS Disposal System Function Specification

DSTS Disposal System Technical Specification

EBS engineered barrier system

GDF geological disposal facility

GDFD Generic Disposal Facility Designs

GTSD Generic Transport System Designs

GWPS Generic Waste Package Specification

HSE Health and Safety Executive

IAEA International Atomic Energy Agency

ILW intermediate level waste

LLW low level waste

LoC Letter of Compliance

LSA low specific activity (material)

MRWS Managing Radioactive Waste Safely

NDA Nuclear Decommissioning Authority

NM nuclear material

OCNS Office for Civil Nuclear Security

ONR Office for Nuclear Regulation

RWMD Radioactive Waste Management Directorate

SAPs Safety Assessment Principles

SCO surface contaminated object

SWL safe working load

SWTC standard waste transport container

WAC waste acceptance criteria

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1 Introduction

In 2001 the UK Government initiated the Managing Radioactive Waste Safely (MRWS) programme with the aim of finding a practicable solution for the long-term management of the UK’s higher activity radioactive wastes. Underpinning this aim was a need to achieve long-term protection of people and the environment in an open and transparent way that was based on sound science and that made effective use of public monies.

In June 2008, in response to recommendations from the Committee on Radioactive Waste Management (CoRWM) [1], the UK Government published the MRWS White Paper [2]. The White Paper confirms the Government’s acceptance of CoRWM’s recommendation that geological disposal is the best available approach for the long-term management of higher activity radioactive wastes. It also sets down the framework by which the geological disposal option is to be implemented.

A key aspect of the MRWS White Paper is that it envisages that the Radioactive Waste Management Directorate (RWMD) of the Nuclear Decommissioning Authority (NDA) will evolve into the organisation responsible for the delivery of a geological disposal facility (GDF).

As implementer and future operator of a GDF, and therefore as the ultimate receiver of wastes for disposal, RWMD will be responsible for the production of waste acceptance criteria (WAC) for such a facility. Whilst plans for the construction of a GDF remain at an early stage, the information necessary to define final WAC is not available. In the meantime, and as a precursor to WAC, we produce packaging specifications, the primary purpose of which is to enable the holders of radioactive wastes to condition those materials in a form that will be compatible with the anticipated needs of transport to and disposal in a GDF.

In 2010 we published proposals for a significant updating of our packaging specifications [3]. These proposals included the development of a hierarchy of packaging specifications defined in such a manner to ensure that the needs of all users would be satisfied. The purpose of this Generic Specification is to apply the high-level packaging requirements defined by the Generic Waste Package Specification (GWPS) [4] to waste packages containing low heat generating wastes1. These requirements are derived from the anticipated needs for the transport and geological disposal of all waste packages containing such wastes. They are applied to specific designs of waste packages, those that would result from the use of standardised designs of waste container which have been shown to be compatible with our current plans for geological disposal, in the form of Waste Package Specifications (WPS).

This Generic Specification makes no assumptions regarding the geographical location of a GDF, the geological environment in which it will be constructed or the specific geological disposal concept which could be adopted. Accordingly the packaging requirements are defined in such a manner as to be bounding of a range of disposal concepts that could be implemented for low heat generating wastes in a number of geological environments that exist at a range of locations throughout the UK.

The remainder of this document is structured in the following manner:

Section 2 provides background information on geological disposal in general, RWMD’s approach to defining the requirements for, and demonstrating the safety of a UK GDF. It also summarises the role played by packaging specifications in assessing the disposability of waste packages.

1 A full description of the wastes covered by this Generic Specification can be found in Section 3.

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Section 3 describes the wastes that are covered by this Generic Specification.

Section 4 outlines the approach adopted in the development of this Generic Specification.

Section 5 defines the bases for the definition of the packaging requirements and the assumptions that are made as part of this process.

Section 6 defines the packaging requirements together with a brief explanation of the basis for their definition.

A glossary of important terms and phrases is presented in Appendix A.

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2 Background

2.1 The geological disposal of radioactive waste

The MRWS White Paper sets out the Government’s framework for the long-term management of the UK’s higher activity waste2, a key aspect of which is ‘geological disposal, coupled with safe and secure interim storage’ of such waste.

Whilst the precise manner in which geological disposal would be implemented in the UK is not yet defined we envisage that any approach to long-term management of higher activity waste would comprise a number of distinct stages. These could include:

the manufacture of passively safe and disposable waste packages;

a period of interim surface storage, usually at the site of waste arising or packaging;

transport of the waste packages to a GDF;

transfer of waste packages underground and emplacement in the disposal facility;

back-filling of the disposal areas; and

eventual sealing and closure of the facility.

The exact nature, timing and duration of each stage would depend on a number of criteria, including the geographical location and host geology of a GDF, as well as the geological disposal concept selected for implementation for each distinct type of waste.

The key aim of all of the geological disposal systems implemented or under development worldwide is the containment and isolation of radionuclides and other hazardous materials associated with the waste, the former being achieved by the use of multiple barriers. The barriers provided by a typical geological disposal system include those provided by the waste package, the immediate surroundings of the waste package3 and the geology surrounding the disposal facility. The effectiveness of the disposal system relies on these barriers working together to ensure that radionuclides, and other hazardous materials associated with the wastes, are sufficiently contained and isolated such that they will not return to the surface at levels that could cause harm to people or the environment.

2.2 The Disposal System Specification

As part of our programme for the implementation of geological disposal in the UK, and to set out a clear definition of the requirements of the disposal system, we have developed the generic Disposal System Specification (DSS). These requirements include regulatory and stakeholder requirements, as well as a consideration of the nature, characteristics and quantities of the wastes that are destined for disposal.

The development of the DSS, and the associated systems designs for the transport of waste packages and their disposal in a GDF, is an iterative process with the assessments of safety, environmental effects and cost. The requirements and constraints it defines are continually refined in light of the results from ongoing programmes of work, including design development, assessments and research. Updating the DSS will take into account

2 The description ‘higher activity waste’ encompasses all wastes and radioactive materials

identified in the MRWS White Paper as being potentially destined for geological disposal. 3 These being referred to collectively as the engineered barrier system (EBS).

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the results from work on the inventory, engineering design, site investigations, safety, environmental and sustainability assessment, consideration of security and safeguards issues, research and public and stakeholder engagement.

The DSS comprises two documents:

The Disposal System Functional Specification (DSFS) [5], the purpose of which is to identify and document the overall objectives of and constraints on the disposal system. It describes the high-level requirements for the disposal system and is in a form suitable for a wide range of stakeholders.

The Disposal System Technical Specification (DSTS) [6] underpins and develops the high-level DSFS by describing in more detail and justifying the requirements for and constraints on the disposal system. It defines the scope and bounds of the engineering design work and provides the designers of the disposal system with the requirements that must be satisfied.

The DSS is a starting point for the development of designs for the geological disposal system, which includes those for the transport of waste to a GDF. The details of these designs, as currently envisaged, can be found in the following documents:

Geological Disposal: Steps towards implementation [7], in which we describe the work that we have undertaken so far, and our future work plans for the delivery of a GDF. Notably, as regards the development of packaging specifications, this document identifies the ‘illustrative’ disposal concepts which could be implemented for the disposal of higher activity wastes in various geological environments in the UK.

The Generic Disposal Facility Designs (GDFD) report [8] which describes the illustrative GDF designs developed for three geological environments, comprising higher strength host rock, lower strength sedimentary host rock and evaporite host rocks. It presents our understanding of how geological disposal could be carried out in a range of different geological environments.

The Generic Transport System Designs (GTSD) report [9], which outlines potential generic transport system designs for moving waste to the disposal facility, by road, rail and/or sea. It summarises the hardware, logistics, operations and cost basis for the generic transport system.

2.3 The Disposal System Safety Case

As a means of presenting the methods, evidence and arguments by which we demonstrate the safety of our plans for geological disposal we have developed a generic Disposal System Safety Case (DSSC) [10]. The DSSC is founded on the generic DSS and systems designs for transport and disposal. It comprises a suite of documents which consider the safety of all aspects of the long-term management of waste following its export from the site of interim storage to a GDF, construction, operation, decommissioning and closure of the disposal facility, and the safety of the disposal facility in the very long term, after it has been sealed and closed. Of direct relevance to the definition of the requirements for waste packages are the three generic safety cases for these three periods of the long-term management of waste, namely:

The generic Transport Safety Case (TSC) [11] which summarises why we have confidence that the system for transporting wastes to a GDF would be safe. It gives an overview of how safety would be demonstrated for individual packages, and a summary of an illustrative safety assessment of the transport system.

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The generic Operational Safety Case (OSC) [12], which presents an illustrative safety case for both normal operation and under fault conditions considering disposal in the different geological environments. It provides a preliminary assessment of operational risk, including construction risk, against regulatory limits and targets.

The generic Environmental Safety Case (ESC) [13], which considers the environmental safety of a GDF during the operational period and after closure of the facility. It explains in principle why we have confidence in the environmental safety of a GDF and our approach to developing the necessary safety case to demonstrate that confidence, with reference to the different geological environments.

The DSSC also summarises the current status of the underlying science base in key areas such as waste package longevity, radionuclide behaviour and criticality safety.

2.4 Specifications for packaged waste

When radioactive waste is disposed of in a GDF it will be required to be compliant with the WAC defined for that facility. The WAC would be expected to be produced by the facility operator, under the oversight and scrutiny of the relevant regulatory authorities. They would be based on the safety cases produced for the operational and post-closure periods of the facility and would reflect the requirements for transport, as appropriate.

In the UK, plans for the geological disposal of higher activity radioactive waste are still at an early stage, so the information necessary to develop final WAC is unavailable. However, in order that wastes can be converted into passively safe and disposable forms, as soon as is reasonably practicable, we produce generic packaging specifications. These specifications define the standard properties and performance requirements for waste packages which are compatible with the anticipated systems and safety cases for transport to and disposal in a GDF.

The packaging specifications are an important part of determining the disposability of waste packages (see Section 2.5) and may be considered to act as the ‘preliminary’ WAC for a future GDF. This approach is consistent with that outlined in guidance produced by the International Atomic Energy Agency (IAEA) [14] and with that adopted in a number of countries worldwide.

Packaging specifications are produced with a number of key purposes in mind, notably:

To support the development of our plans for the implementation of geological disposal for higher activity radioactive waste;

To provide the UK nuclear industry and regulators with a clear definition of the requirements for packaged waste in advance of the construction of a GDF;

To provide a basis for the assessment of the suitability of plans to package waste to produce disposable waste packages and thereby permit the early packaging of waste; and

To permit scrutiny of this aspect of our plans to implement geological disposal for higher activity waste in the UK.

2.5 Assessing the disposability of waste packages

The ‘Letter of Compliance’ Disposability Assessment Process has been established as a means of supporting the UK nuclear industry’s ongoing work on the conditioning and packaging of higher activity waste for disposal. The process has been extensively developed over a period of more than 20 years in cooperation with the site operators and

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industry regulators, and in a manner that aligns with regulatory expectations for the long term management of higher activity waste [15].

The main purposes of the Disposability Assessment Process [16] are to:

Give confidence to site operators that the implementation of their proposals to package waste will result in waste packages that meet the anticipated needs for transport to and disposal in a GDF;

Aid in the identification of optimised solutions to the packaging of specific types of waste;

Provide us with confidence that the geological disposal concepts considered within the DSS and DSSC will be appropriate for the wastes they are expected to accommodate; and

Permit the identification of wastes and proposed approaches to packaging that could challenge current disposal concepts and thereby allow early consideration of what changes may be required to those concepts to permit the resulting waste packages to be accommodated.

In the event that a disposability assessment identifies no significant uncertainties in the ability of the proposed packaging approach to produce disposable waste packages, a Letter of Compliance (LoC) can be issued to endorse it, and the waste packages that would result from its implementation. LoCs can be issued at one or more of a number of key stages during the development of plans to package waste, and at which a disposability assessment takes place. Depending on the stage of the development of the packaging process, the issue of a LoC indicates:

Conceptual stage LoC: That the proposed waste package would in principle be compliant with the generic geological disposal concept(s).

Interim stage LoC: That evidence has shown that the as-designed waste package would be compliant with the generic geological disposal concept(s).

Final stage LoC: That evidence has shown that the as-manufactured waste package will be compliant with the generic geological disposal concept(s).

Regulatory guidance [15] suggests that, as part of plans to retrieve and package a specific waste stream, site operators should produce a radioactive waste management case (RWMC) which would include arguments as to the disposability of the waste packages that would result from their plans. The outcomes of a disposability assessment and, in particular, the safety arguments that underpin any endorsement of the proposed waste packages by way of the issue of a LoC, provide key inputs to the development of a RWMC.

As discussed in Section 4 the disposability of the waste packages that would result from the implementation of a proposed approach to the packaging of waste is based on their compliance with the relevant packaging specification. In general a Conceptual stage disposability assessment will be carried out against the relevant Generic Specification for the type of waste to be packaged, whereas Interim and Final stage assessments will be carried out against the WPS for the waste packages that would result from the use of a specific design of waste container.

The outcome of a disposability assessment may include the identification of some aspect of the proposed waste packages not being compliant with the relevant packaging specification, and thereby potentially with the geological disposal concept. In such an event a LoC would not be issued but it may be that a case will exist for some aspect of the geological disposal concept to be changed to accommodate the proposed waste packages. We have established a concept change control management process to ensure that

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proposed changes to the geological disposal concept are recorded, evaluated and, if appropriate, implemented at an appropriate time and in a consistent way. This includes ensuring that changes are recorded in the relevant documentation, such as the DSS, the disposal systems designs and the relevant packaging specifications.

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3 The wastes covered by this Generic Specification

This Generic Specification defines the packaging requirements that are applicable to all waste packages containing a broad range of wastes which are described as having ‘low heat generation’, this broad description being intended to cover wastes which have similar radiological properties to intermediate level waste (ILW). As a consequence, all such wastes could be subject to geological disposal utilising disposal concepts which we have identified as being potentially suitable for implementation in the UK for the geological disposal of ILW [7].

ILW is defined in a number of sources (e.g. [2]) as:

‘Radioactive wastes exceeding the upper activity boundaries for low level waste (LLW) but which do not need heat to be taken into account in the design of storage or disposal facilities.’

All radioactive waste produces radiogenic heat from the radioactive decay of the radionuclides associated with them. The radiogenic heat output of wastes classed as ILW is generally low in the conventional sense, the average heat output of all of the ILW waste streams recorded in the UK Radioactive Waste Inventory (UKRWI) [17], when conditioned for disposal, being ~1Wm-3. In most contexts such a heat output would be considered low, however heat output does have to be considered for the design of transport and GDF systems where various regulatory and operational constraints on temperature will apply. This is reinforced by the fact that the average heat outputs of some ILW waste streams are up to two orders of magnitude higher than the overall average value for ILW and there is also significant variation within some waste streams which could result in even higher radiogenic heat outputs for individual waste packages.

Whilst radiogenic heat output is not the only characteristic of waste that has to be considered in the design of a geological disposal system, it is a useful discriminator when identifying a broader range of wastes that could be disposed of in accordance with a particular disposal concept. There are a number of other wastes and potential wastes which have radiogenic heat outputs which lie within the range of that for the ILW in the UKRWI. Accordingly, waste packages containing such materials could potentially be safely managed by way of the same disposal concepts as those identified as suitable for ILW. These potentially include wastes containing separated plutonium4, bulk uranium5 and also unirradiated or lightly irradiated reactor fuel (e.g. from research reactors). Additionally there are some types of LLW that are not suitable for near surface disposal6 (generally as a result of the presence of long-lived radionuclides) and for which geological disposal is a safe option for their long-term management.

As stated above, radiogenic heat output is not the only discriminator for what types of waste could be considered sufficiently similar to ILW to permit geological disposal in accordance with the same concepts. The presence of significant quantities or concentrations of radionuclides with specific properties in a waste may mean that some disposal concepts developed for ILW may not deliver the required level of safety for such wastes. This could include:

4 It should be noted that not all of the UK’s stocks of separated plutonium may be suitable for

disposal as ‘low heat generating waste’. 5 Including uranic materials in a range of physical and chemical forms, and uranium-235

concentrations, which have arisen from a variety of nuclear fuel cycle activities. 6 These are, for example, those types of LLW that would not be compliant with the WAC for the

Low Level Waste Repository (LLWR) in Cumbria.

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radionuclides of significance for criticality safety (i.e. fissile isotopes of uranium and plutonium); and

long-lived radionuclides of significance for the post-closure safety of a GDF.

A decision as to the suitability of a specific waste for geological disposal in accordance with concepts defined for ILW would normally be made at an early stage in the Disposability Assessment Process.

In summary this Generic Specification can be applied to waste packages containing:

all wastes identified as ILW in the UKRWI;

any LLW that is not suitable for near surface disposal; and

any other radioactive material for which the concepts identified as suitable for the disposal of ILW [7] could be shown to provide an appropriately safe approach to their long-term management.

It should be noted that whilst geological disposal concepts may be developed for materials such as bulk uranium and separated plutonium, and Generic Specifications developed for these concepts, this would not preclude the use of this Generic Specification as the basis for the performance of Conceptual stage disposability assessments for waste packages containing such materials.

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4 The development of this Generic Specification

This Generic Specification is aimed at waste owners and/or packagers, industry regulators and all stakeholders with an interest in how packaging standards are defined for radioactive wastes generally characterised by having low heat generation. It has been produced particularly to facilitate the development of new approaches to the packaging of those types of wastes, using new waste container designs and/or waste conditioning processes.

The form of the packaging requirements are such as to permit waste packagers to develop packaging approaches that are proportionate to the hazard presented by particular wastes whilst ensuring that waste packages perform in a manner that will ensure safe transport to and disposal in a GDF. This is achieved by defining what is required of waste packages, in terms of their properties and performance, rather than how that is to be achieved.

The general form of this Generic Specification is therefore to set down bounding values for the standard properties of waste packages (e.g. maximum dimensional envelope, maximum gross mass) together with their performance requirements (e.g. waste container durability, waste package accident performance).

The GWPS [4] is the document that defines high-level requirements for all waste packages that are destined for geological disposal. By defining the safety functions that are required of all waste packages in order for them to be compatible with the needs of safe transport to and disposal in a GDF, the GWPS defines a series of packaging criteria and accompanying high level packaging requirements.

In this Generic Specification the packaging requirements defined by the GWPS are used as the basis for the definition of more quantitative requirements that apply to waste packages containing low heat generating waste. As part of this process information from a number of documents has been used, including the generic DSS (notably the DSTS [6]), the transport and disposal facility design reports [8, 9] and the generic DSSC, notably the three safety cases identified in Section 2.3; the generic TSC [11], OSC [12] and ESC [13].

These documents, in particular the generic DSS, are used to ensure that the packaging requirements defined by this Generic Specification are both bounding of all of the requirements of transport and geological disposal and proportionate to the hazards presented by the waste to be packaged. This notably includes the regulatory framework applying to transport and disposal, and information regarding the physical, chemical and radiological properties of the waste to be packaged. The development of this Generic Specification has also drawn on experience gained by other waste management organisations, in the UK and overseas, in the development of packaging specifications and/or WAC for a range of different categories of radioactive waste and disposal concepts.

A key role for this Generic Specification is as a baseline for the Conceptual stage disposability assessment of proposals to package low heat generating wastes, as defined in Section 3, for geological disposal. Accordingly, the issue of a Conceptual stage LoC for such proposals will require the demonstration of the compliance of the proposed waste packages with the packaging requirements contained herein.

This Generic Specification also acts as the basis for the definition of the WPS which define the requirements for waste packages containing low heat generating waste that would result from the use of standardised designs of waste container. WPS are produced for designs of waste container which have been shown to be compatible with our current understanding of the needs of the transport to and disposal of waste packages in a GDF. Further WPS will be added as and when additional designs of waste container are identified and shown to be similarly compliant, and specifically with the requirements defined by this Generic Specification. The issue of an Interim or Final stage LoC will require a waste package to be shown to be fully compliant with a specific WPS.

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As a means of making the full range of the RWMD packaging specifications readily available to waste producers and other stakeholders a suite of documentation known as the Waste Package Specification and Guidance Documentation (WPSGD) is published and maintained for ready access (i.e. via the NDA website, www.nda.gov.uk). The WPSGD comprises:

The GWPS [4];

All published Generic Specifications;

All published WPS;

Guidance on the achievement of the requirements defined by the WPS;

Guidance on the achievement of the requirements for wasteforms;

Specifications for quality management and data recording requirements during waste packaging and interim storage, together with supporting guidance; and

Thematic guidance on a range of topics related to the packaging of waste with specific properties.

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5 Basis for the definition of the packaging requirements

The fundamental aims for the packaging of waste are to ensure that the resulting waste packages are:

passively safe and suitably robust physically, so as to ensure containment and safe handling during all ensuring phases of the long-term management of the waste including disposal at a GDF;

suitable for safe transport through the public domain in compliance with the relevant regulations for such transport; and

compatible with the safety cases for the operational and post-closure periods of a GDF.

The waste package provides the most immediate barrier to the release of radionuclides and other hazardous materials from the waste it contains, both during interim storage, transport and within a multiple barrier geological disposal system. It also plays a key role in protecting individuals from the radiation emitted by the radionuclides it contains during interim storage, transport and the GDF operational period.

The barrier provided by a waste package can be considered to comprise two components, each of which can act as a barrier in its own right:

The waste container, which provides a physical barrier and also enables the waste to be handled safely during and following waste package manufacture. Containers can be manufactured from a range of materials with designs selected to suit the requirements for the packaging, transport and disposal of the wastes they contain.

The wasteform, which can be designed to provide a significant degree of physical and/or chemical containment of the radionuclides and other hazardous materials associated with the waste. The wasteform may comprise waste which has been ‘immobilised’ (e.g. by the use of an encapsulating medium such as cement) or that which may have received more limited pre-treatment prior to packaging (e.g. size reduction and/or drying).

Both the waste container and the wasteform therefore contribute to the achievement of the required performance of the waste packages, the relative importance of each generally depending on the robustness of the former. This is illustrated in Figure 1 which shows in stylised form how the use of a more robust waste container can reduce the required contribution of the wasteform to overall waste package performance.

Figure 1 also shows that for all waste packages both the waste container and the wasteform will be required to play some role. It should also be noted that it is the overall performance of the waste package, rather than that of its two components, that is the governing factor in judging its disposability.

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Figure 1 Relative contribution of the waste container and wasteform to waste package performance

This Generic Specification is founded on the requirements for geological disposal defined by the DSS and the high-level requirements for all waste packages as defined by the GWPS [4]. The former includes a requirement that all waste packages should be capable of being safely transported to a GDF in accordance with the systems defined by the GTSD [9] and, following receipt at a GDF, of being safely handled by way of the processes and equipment defined in the GDFD [8]. Also included is a consideration of the required performance of waste packages in the GDF post-closure period, as defined by the DSTS in the form of a series of post-closure safety functions.

No explicit consideration of the needs of interim storage has been included in the definition of the packaging requirements that make up this Generic Specification, as waste packages will be designed by the site operator to be compatible with needs of their storage facilities.

Historically, the packaging specifications for ILW (e.g. the 2007 GWPS [18]) have tended to be based on the use of a limited number of standardised waste containers and transport containers. The generic DSS and the transport and GDF systems designs identify the types of waste containers that could be used for the packaging of ILW, which include those identified in the 2007 GWPS7. Whilst these containers are suitable for use for the packaging of the wastes to which this Generic Specification applies, they should not be seen as the only designs that could be used. Other designs of waste container could be added if they can be shown to be compatible with the geological disposal concept, by way of the concept change control management process (see Section 2.5). If this can be shown to be the case, the waste container can be added to those identified by the DSS and a WPS, based on the requirements defined by this Generic Specification, produced for the waste packages that it could be used to manufacture.

7 These being the 500 litre drum, the 3 cubic metre box and drum, and the 2 and 4 metre boxes,

together with the container used for the storage of waste in the Miscellaneous Beta Gamma Waste Store at Sellafield and that used for the packaging of wastes arising from the decommissioning of the Windscale Advanced Gas-cooled Reactor.

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5.1 The transport of waste packages to a GDF

In the absence of a geographical location for a GDF, the TSC currently assumes that all of the waste packages stored at the several nuclear sites throughout the UK will have to be transported, through the public domain, to a GDF.

The transport of radioactive materials is subject to a number of requirements implemented into UK law8, notably the IAEA Regulations for the Safe Transport of Radioactive Material9 [19]. The IAEA Transport Regulations define general requirements and, in some cases, quantified limits for a range of properties of radioactive materials which apply to their transport and these are, where relevant, incorporated into this Generic Specification.

With regard to transport the distinction between a ‘waste package’ and a ‘transport package’ is important as it influences the manner by which the requirements of the IAEA Transport Regulations are applied to waste packages. A waste package will, in general, comprise a container in which waste is placed and which is suitable for disposal without further treatment. Some waste packages may be capable of satisfying the requirements of the IAEA Transport Regulations, without additional protection for transport, and are described as ‘transport packages in their own right’. The requirements of the IAEA Transport Regulations are therefore applied directly to such waste packages.

Some designs of waste package might not be suitable for transport without additional protection to provide, for example, radiation shielding and/or physical containment of its contents. In such cases a transport package may comprise a reusable transport container into which one or more waste packages are placed. On receipt at a GDF the waste packages would normally be removed from the transport container prior to disposal. For such waste packages significant benefit can be claimed for the protection that is provided by the transport container when the requirements of the IAEA Transport Regulations are considered, and this is reflected in the application of these requirements to the waste packages themselves.

The constraints of the transport system and the regulations which apply to the transport of waste packages through the public domain are in many cases the most limiting in the definition of the packaging requirements that make up this Generic Specification. It is therefore important that the assumptions regarding the transport of waste packages are clearly defined. At the highest level this includes an assumption that all waste packages will be transported using the systems and operational procedures defined by the GTSD [9].

The IAEA Transport Regulations define two regimes under which transport packages can be carried, these being under the conditions defined as ‘exclusive use ‘ and ‘non-exclusive use’. Paragraph 221 defines ‘exclusive use’ as:

‘…the sole use, by a single consignor, of a conveyance or of a large freight container, in respect of which all initial, intermediate and final loading and unloading is carried out in accordance with the directions of the consignor or consignee’.

If all of these provisions do not apply to a transport operation it is deemed to take place under the conditions of non-exclusive use. The controls on several aspects of transport package performance (e.g. external dose rate, surface temperature, quantities of fissile material) are less onerous if transport takes place under the more restrictive operational conditions of exclusive use. In line with a conservative approach, the packaging

8 The Radioactive Materials Transport Team of the ONR has regulatory responsibility for the

transportation of radioactive material in Great Britain. 9 This reference will be referred to as the ‘IAEA Transport Regulations’ in the remainder of

document, and direct reference made to relevant Paragraphs in those Regulations.

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requirements that make up this Generic Specification are generally based on transport taking place under the conditions of non-exclusive use and the more onerous controls applied where relevant.

The IAEA Transport Regulations define a number of categories of transport package, two of which are of relevance to this Generic Specification:

Type B, for the transport of much larger quantities of activity, and for which protection of transport workers and members of the public relies on the structure of the transport package itself; and

Industrial Packages (Type IP), for the transport of materials with limited specific activity and for which protection is vested in controls placed on the form and specific activity of the contents of the transport package.

Two classes of Type B transport package are defined:

Type B(U) - Approved for use anywhere in the world after approval by a competent authority10 in a single country;

Type B(M) - Approved for use in a specified country, or specified group of countries, after approval by a competent authority in each country.

The controls defined for the transport of the two classes of Type B transport package are different, with those for Type B(U) transport packages being more onerous. In line with a conservative approach for the definition of requirements for waste packages, the conditions for Type B(U) transport packages are generally used as the basis for the derivation of the requirements in this Generic Specification for all waste packages transported as, or as part of, Type B transport packages. It should however be noted that, at the discretion of the competent authority, the controls defined for Type B(M) transport packages could be applied.

Type IP transport packages are defined at three levels, IP-1, IP-2 and IP-3, on the basis of the physical form and allowed specific activity of their contents and the operational controls required during their transport (i.e. whether they are transported under the conditions of exclusive or non-exclusive use). The specific activity of the contents of all Type IP transport packages are limited to those defined by the IAEA Transport Regulations as low specific activity (LSA) material or surface contaminated objects (SCOs). In general terms:

Type IP-1 transport packages are only permitted to carry LSA material and SCOs with very low specific activity (LSA-I and SCO-I). Wastes with such low specific activities are unlikely to be considered for geological disposal as disposal at a near surface facility (e.g. the LLWR) would be deemed a more appropriate approach for their long-term management.

Type IP-2 transport packages can be used for the transport of solid LSA material and SCOs with significantly higher specific activities although for LSA material with the highest specific activity (i.e. LSA-III material), the conditions of exclusive use will be necessary for transport.

Type IP-3 transport packages can be used to transport all types of LSA materials and SCOs, the former being permitted to be in liquid or gaseous form.

The TSC assumes that waste package transported as, or as part of, Type IP transport packages will be subject to the specific controls defined for Type IP-2 transport packages.

10 The competent authority for the transport of radioactive material in Great Britain is the Secretary

of State for Energy and Climate Change.

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5.2 The disposal of waste packages in a GDF

For a waste package to be deemed ‘disposable’ it must be both physically compatible with the systems defined for transport and the GDF, and with the assumptions that underpin the safety cases for transport and the operational and post-closure periods of a GDF. The design of a GDF, which will be strongly influenced by the geological environment in which it is constructed, will therefore place significant constraints on the properties and performance of waste packages.

A wide range of different geological environments that could be suitable for hosting a GDF for higher-activity radioactive wastes exist in the UK. However, at the current stage in the MRWS process, no site for a GDF has been selected and so the actual design of a GDF has not been defined. It is therefore necessary that the packaging requirements in this Generic Specification are defined in such a manner that will bound a sufficiently representative range of GDF designs.

Drawing from work to investigate geological disposal concepts that have been planned or implemented worldwide [20], the DSS identifies an ‘illustrative’ concept for the geological disposal of ILW which could be implemented in each of three generic geological environments11 in which a GDF could be constructed in UK [7]. The demands imposed on waste packages by the GDF designs that would result from the implementation of these three illustrative concepts could result in a need for different packaging requirements to be defined.

The generic packaging requirements defined by the 2007 GWPS [18] were based on an assumption that a GDF would be based on the Phased Geological Repository Concept (PGRC) [21] constructed in a geological environment defined as ‘higher strength rock’. Such an approach was considered to result in the definition of packaging requirements that were bounding of other approaches to geological disposal and that would not unduly preclude the adoption other viable approaches to the long term management of waste packages manufactured to comply with them. During the development of this Generic Specification we commissioned work to consider the implications for the definition of packaging requirements of the adoption of either of the illustrative concepts as a means of ensuring that those defined herein are bounding of all three concepts [22]. The outcomes of this work have been incorporated into the packaging requirements defined in Section 6.

11 These being described as ‘higher strength rock’, ‘lower strength sedimentary rock’ and

‘evaporites’.

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6 Requirements for waste packages containing low heat generating waste

This Section defines the requirements for all waste packages containing wastes that fall into the category of low heat generating waste, as defined in Section 3, that take into account the needs for the transport of those waste packages and their disposal in a GDF. The requirements are defined for each of the packaging criteria identified in the GWPS [4] and are an application of the high-level requirements defined therein, to waste packages containing low heat generating waste.

The packaging requirements specified below are, in general, defined for the complete waste package but in practice the manner in which they are achieved will depend on a number of factors including:

the nature of the waste container;

the physical, chemical and radiological properties of the waste; and

the means by which the waste is conditioned for disposal.

Accordingly, to aid the use of the packaging requirements in the development of plans to package waste, they are grouped in a manner to reflect those which are most directly related to the waste container, the wasteform, and the waste package as a whole.

In addition, a number of requirements are defined for the controls that will need to be applied during the manufacture and storage of waste packages.

It should be noted that, where the words ‘shall’ and ‘should’ are used in the packaging requirements, their use is consistent with the recommendations of BS 7373:1998 [23] in that they have the following meaning:

‘shall’ denotes a limit which is derived from consideration of a regulatory requirement and/or from a fundamental assumption regarding the current designs of the transport or disposal facility systems;

‘should’ denotes a target from which relaxations may be possible if they can be shown12 not to result in any significant reduction in the overall safety of the geological disposal system.

A number of the packaging requirements (e.g. heat output, gas generation and criticality safety) include quantified ‘screening levels’. These values are defined to provide guidance to waste package designers by indicating which proposed waste packages would be expected to satisfy the relevant requirement without additional justification. They also act as a ‘flag’ to indicate those waste packages for which such justification will be required as part of a submission for the assessments of their disposability. It should be noted that these screening levels are not intended to be used as a sole basis for the development of packaging proposals as, in many cases, the actual limiting values for waste packages may be significantly higher.

The format of this Section is to define each packaging requirement in terms of that defined by the GWPS (shown in bold italic type) together with any additional requirements for the waste packages that are the subject of this Generic Specification (shown in bold type).

12 This would generally be by way of the Disposability Assessment Process.

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Each packaging requirement is followed by a brief explanation of the basis for its definition. A more detailed explanation, together with a justification of the quantified limits for specific designs of waste package can be found in the guidance material that supports the WPS.

6.1 Requirements for waste containers

The properties of the waste container shall be such that, in conjunction with those of the wasteform, it satisfies all of the requirements for the waste package.

In Section 5 the contribution that the waste container can make to the overall properties and performance of a waste package was discussed. For some of the required waste package properties (e.g. external dimensions, lifting features, and identification) the waste container will generally satisfy the requirement, whereas for others (e.g. stackability, accident performance) it may only play a partial role, the actual extent of the role played by the waste container depending on its robustness, as illustrated by Figure 1. It should however be noted that, whilst some designs of waste container will be able to provide most aspects of required waste package performance, this will not preclude the need for the wasteform to satisfy the requirements defined in Section 6.2.

In general terms it is expected that the waste container will be required to provide the waste package with adequate:

mechanical strength to:

o withstand stacking forces (Section 6.1.3);

o resist damage due to pressurisation by internally generated gases (Section 6.3.6);

o ensure that the specified impact accident performance (Section 6.3.8) can be achieved; and

o withstand other loads that may occur during the long-term management of the waste package, as required by the ESC.

radiation shielding13 to ensure that the external dose rate is minimised and that specified limits are not exceeded (Section 6.3.3);

thermal properties to ensure that the required fire accident performance (Section 6.3.8) and other thermal requirements of the waste package will be achieved; and

resistance to degradation to ensure the overall integrity of the waste container is maintained for an appropriate period (Section 6.1.5).

6.1.1 External dimensions

The external dimensions of the waste package shall be compatible with the transport and GDF handling systems.

The overall dimensions of a transport package should not exceed 6.058m x 2.438m plan x 2.591m high.

The dimensions of a transport package carried by rail shall not exceed 2.67m wide or 2.40m high.

13 It should be noted that some designs of waste container will provide little radiation shielding and,

where required, this will be provided by the use of a transport container or similar device.

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The external dimensions of waste packages (and transport packages) are limited by the handling systems used during transport and at a GDF.

The current GDF handling systems are based on transport packages having dimensions that lie within the envelope defined for Series 1 freight containers as specified in ISO668 [24].

Waste packages could be transported to a GDF by road, rail, sea or inland waterway, or by a combination of these means. In general, transport by rail is the most restrictive from the point of view of the external dimensions of waste packages. The maximum overall dimensions of a transport package must be compatible with the dimensions envelope defined by Standard W6A Rail Gauge [25]. This leads to a maximum transport package width of 2.67m14 and a maximum height of 2.40m [26]. Transport by rail will also impose limits on the length of transport packages although these will be generally less restrictive than the GDF limit (i.e. ~6m). Less restrictive rail gauges exist although this could limit which parts of the network could be used. Larger waste packages could be transported by road; although the transport of larger waste packages in this manner may ultimately be limited by restrictions on their gross mass (see Section 6.3.2).

The limiting dimensions stated above apply directly to waste packages which are transport packages in their own right. For waste packages requiring additional protection during transport they will apply to the external dimensions of the transport container. Three designs of standard waste transport container (SWTC) are currently under development:

The SWTC-15015 has the largest cavity size and can accommodate waste packages with external dimensions of up to 1.85m plan by 1.37m high.

For waste packages requiring greater radiation shielding the SWTC-28515 would have to be used, and can accommodate waste packages with external dimensions of up to 1.72m plan by 1.245m high.

The more lightly shielded SWTC-7015 which, by virtue of its lower unladen weight, can be used for road transport, and which has the same cavity size as the SWTC-285.

6.1.2 Handling feature

The waste package shall enable safe handling by way of the transport and GDF handling systems.

The waste package shall incorporate handling features to enable lifting under a load equivalent to twice the maximum specified gross mass without any effect that would render it non-compliant with any of the requirements defined in this Specification.

Where tie down within a conveyance is necessary for their safe transport, waste packages which are transport packages in their own right shall incorporate tie-down features suitable for their maximum specified gross mass.

The design of the waste package should enable remote handling.

To permit their safe and efficient handling waste packages should incorporate handling features designed in such a manner as to be compatible with the handling systems that are assumed in the GTSD and GDFD [8, 9]. Designs which are not so compliant may also

14 These values represent the total envelope permitted by the W6A rail gauge and therefore the

effects of any peripheral equipment (e.g. weatherproofing covers, impact limiters etc.) used during transport would have to be taken into account.

15 The numeric part of the identifier indicates the nominal thickness of stainless steel shielding provided by the transport container.

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satisfy the generic requirements for handling but this would need to be considered as part of a disposability assessment and evaluated by way of the concept change control management process.

These handling features must be able to withstand the full range of forces which could be applied during all normal waste package handling operations. This includes a requirement for them to be able to withstand the loads that would result from the lifting of a waste package with twice the specified maximum gross mass, to take into account of the so-called ‘snatch factor’ [27].

Waste packages which are transport packages in their own right will need to be restrained during transport and this can be achieved by the inclusion of ‘tie-down’ features in the waste container design. This is an explicit requirement for waste containers which are classed as ISO freight containers, and which have followed that route for their approval for the transport of radioactive material.

Whilst the GDFD assumes that some ILW waste packages will be contact handled, in order to retain flexibility in the definition of operational procedures at a GDF (and during transport), and to help ensure that operator doses will be as low as reasonably practicable (ALARP), the option for their use with remote handling equipment should be retained in the design of the handling features of all waste packages.

For some waste package designs the handling feature will also play a significant role in their safe stacking (Section 6.1.3).

6.1.3 Stackability

Where required by the transport or disposal system, the waste package shall enable safe stacking.

Waste packages which rely on their design to withstand stacking loads should be capable of being stacked to a height of 11m with other waste packages of the same design, each with their maximum specified gross mass. This loading shall not result in any effect that could render the waste package non-compliant with any of the requirements defined in this Specification.

Waste packages which are transport packages in their own right shall comply with the stacking requirements defined by the IAEA Transport Regulations.

Many of the designs of waste package that are covered by this Generic Specification will be stacked in the GDF disposal vaults, although some designs may make use of devices such as stillages for that purpose. Waste packages which do not rely on such devices will need to be capable of being stacked, as would be required by the design of a GDF, and without suffering any effects that could threaten their safe onward management.

The height to which waste packages will be stacked will depend on the design of the GDF disposal vaults, the geological environment in which they are constructed and on the nature of the waste package. The GDFD report makes assumptions with regard to maximum stack heights which will be governed by the ability to excavate vaults of a given height in the ultimately selected host geology, whilst leaving sufficient clearance for any emplacement equipment (e.g. overhead cranes) [8]. This results in a maximum stack height of 11m for waste packages stacked using a stacker truck and 8.7m using an overhead crane.

The IAEA Transport Regulations (Paragraph 723) specify a ‘stacking test’ as part of the process for demonstrating the ability of transport packages16 to withstand normal conditions of transport. This requires transport packages to be capable of withstanding a compressive

16 This requirement does not apply to transport packages whose shape precludes stacking.

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vertical load equal to five times their maximum weight, following which it must be capable of preventing the loss or dispersal of its contents or of its shielding integrity (Paragraph 622). This requirement applies to all waste packages which are transport packages in their own right.

6.1.4 Identification

The waste package shall enable unique identification until the end of the GDF operational period.

The waste package shall be marked at multiple defined locations with a unique alpha-numeric identifier.

The waste package shall remain identifiable by automated systems for a minimum period of 150 years following manufacture.

The application of a unique identifier enables the identification and tracking of every waste package throughout the different stages of its long-term management, and helps to ensure the permanent assignment of the appropriate data record to that waste package.

The use of a waste package identification system based on alpha-numeric identifiers ensures the maximum flexibility and capacity of a system which will need to be capable of ensuring that in excess of 105 waste packages, arising from multiple sites and packaging plants, will be handled safely and efficiently following receipt at a GDF.

For automated reading systems to operate effectively, multiple standardised locations will be specified for identifiers. This will aid in the ease of reading by reducing the need for the waste package to be moved to facilitate identification or of identifiers being obscured by handling equipment. It also provides redundancy in the event of damage to individual identifiers (for example that caused by corrosion) and will reduce the risk of waste packages becoming unidentifiable. Identifier locations are selected to ensure that the application of the identifiers does not comprise the durability of the integrity of the waste container (Section 6.1.5). This latter aspect is also a consideration when the method of applying identifiers is selected.

Making the identifier ‘machine-readable’ and the use of a format containing check digits allows the waste package to be identified remotely by automated systems and the veracity of its identifier confirmed. The use of a standard character set, such as OCR-A characters [28], of a specified size permits waste package identification by either automated or direct visual (i.e. by human operators) means.

A waste package will need to remain identifiable at least until the time at which it is surrounded by the backfill material or when the disposal area has been closed. On the basis of current planning assumptions regarding when a GDF would be available to receive waste packages for disposal (i.e. 2040) and the anticipated length of the GDF operational period (assumed in the GDFD to be ~100 years [8], a minimum period of 150 years is defined as that for which waste package identification by way of its identifiers should be maintained.

Waste packages which are transport packages in their own right will also be required to bear identifiers in accordance with the regulatory requirements for their transport.

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6.1.5 Durability of waste container integrity

The waste package shall enable safe handling by way of its handling feature until the end of the GDF operational period.

The waste container shall maintain containment for as long as is required by the GDF safety case.

The integrity of the waste container should be maintained for a period of 500 years following manufacture of the waste package.

Integrity is defined as the ability of a waste container to provide containment of its contents and of a waste package to be safely handled and stacked. Two of the five operational safety functions defined for waste packages in the GWPS, for containment and safe handling, rely heavily on the maintenance of waste container integrity for an appropriate period. Other safety functions, notably the ability of the waste package to withstand internal and external loads, will also rely at least in part on such integrity. The requirement for the durability of waste container integrity is therefore defined in terms of the period for which the waste container needs to maintain the containment of its contents, the surety of its handling features and its ability to withstand all anticipated loads, notably those resulting from stacking.

Regulatory guidance on the conditioning and disposability of higher activity waste states that ‘A minimum package lifetime of 150 years should be set for design purposes’ [29]. Such a period broadly aligns with current planning assumptions regarding when a GDF would be available to receive waste packages for disposal (i.e. 2040) and the anticipated length of the GDF operational period (assumed in the GDFD to be ~100 years [8]).

The potential for retrieval of waste packages from the disposal vaults in the event of the GDF operational period extending beyond 2140 must also be taken into account when defining the period over which the integrity of the waste container is required to be maintained. RWMD’s position on retrievability, which is consistent with that stated in the MRWS White Paper, is that activities concerned with the development and implementation of geological disposal will be carried out in such a way that the option of retrievability is not excluded [30].

The DSTS defines a number of safety functions that the waste container (and the wasteform, see Section 6.2) will be required to provide for waste packages in the post-closure period. These include preventing groundwater from reaching the wasteform, which is identified as a means of delaying the release of radionuclides into the other components of the EBS. In the case of waste packages containing ILW this refers notably to the retention of relatively short-lived water soluble radionuclides (e.g. strontium-90 and caesium-137, each with half-lives of ~30 years). Whilst indefinite retention of such radionuclides is not the aim, the waste container should provide an effective barrier for a period that would permit them to decay to relatively insignificant levels before their release.

The presence of an engineered vent, to prevent excessive pressurisation of waste packages by internally generated gases, could be viewed as conflicting with the requirements for waste container integrity. In such cases the design of the vent should be such as to minimise ingress of groundwater into the waste package in the post-closure period, such that the flow of water through the waste package would be negligible and that activity would only released in relatively small quantities, mainly by diffusion through the vent.

In order to satisfy the potential requirements of both the operational and early post-closure periods, the need to maintain waste container integrity for 500 years, as specified in the 2007 GWPS [18], has been retained. We have carried out work which shows that current designs of waste container, designed to meet the durability requirement identified by regulatory guidance (i.e. 150 years), would also be expected to maintain an appropriate

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level of integrity for at least 500 years [31]. Notwithstanding this we acknowledge that after 150 years waste packages may need to be handled by means which do not involve the use of the integral handling feature.

The ability of a specific design of waste container to meet this durability requirement will be assessed by way of the Disposability Assessment Process (see Section 2.5) which, as well as considering the design of the waste container itself, will also take into account the potential consequences of the contents of the waste package for the durability of waste container integrity.

6.2 Requirements for wasteforms

The properties of the wasteform shall be such that, in conjunction with those of the waste container, it satisfies all of the requirements for the waste package.

The properties of the wasteform shall comply with the requirements for containment within the geological disposal concept, as defined by the GDF safety case.

The physical, chemical, biological and radiological properties of the wasteform shall:

make an appropriate contribution to the overall performance of the waste package; and

have no significant deleterious effect on the performance of the waste container.

Evolution of the wasteform shall ensure maintenance of the waste package properties that are necessary for safe transport and operations at a GDF.

Evolution of the wasteform shall ensure maintenance of the required safety functions for post-closure performance as set out in the ESC.

As discussed in Section 5 the required performance of a waste package will be provided by a combination of the properties of the waste container and the wasteform it contains. Waste packages manufactured using thin-walled waste containers may rely to a significant degree on the properties of the wasteform if they are to perform in an appropriate manner. By contrast, when a more robust waste container is used, the wasteform will play a lesser role although this will not completely preclude the need for it to possess the properties required to ensure appropriate waste package performance.

The properties of the wasteform will play a key part in ensuring the passive safety of a waste package, irrespective of the nature of the waste container. Wastes should therefore be conditioned to minimise chemical reactivity and to satisfy some basic requirements as to their physical and biological properties. This should extend to ensuring the compatibility of the wasteform and the material from which the waste container is fabricated. These general requirements for wasteforms can be achieved by sorting, segregation and/or a range of pre-treatment processes to ensure the appropriate control of the quantities of some types of material, or of wasteform properties, that could affect the overall performance of the waste package or the other barriers that make up the geological disposal system. Typically this could include controls on the presence of:

free liquids;

activity or hazardous materials in particulate form;

voidage;

in-homogeneity;

reactive materials;

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other hazardous materials17; and

materials that could have a deleterious effect on the other barriers that make up the geological disposal system.

The extent of such controls will be very dependent on the robustness of the waste container and the consequences of the presence of these materials and wasteform properties for waste package and disposal system performance. This would normally be assessed as part of the disposability assessment of a proposed waste package design.

In relation to the performance of the waste container consideration should be given to the potential for chemical reactions between the wasteform and the inner surfaces of the waste container or expansive corrosion of components of the waste that could result in forces being exerted on the waste container.

Evolution of the wasteform, resulting from chemical, biological and/or radiation induced processes will change the properties of the wasteform with time. It is important that such evolution will not result in changes that render the waste package incompatible with the needs of transport or the requirements for safety in the GDF operational period.

In the post-closure period the wasteform may continue to play a role in the overall safety of a GDF. The DSTS defines a single post-closure safety function for wasteforms requiring them to ‘provide a stable, low-solubility matrix that limits the rate of release of the majority of radionuclides by dissolution in groundwater that comes into contact with the wasteform’. Accordingly the consequences of evolution should be such that this requirement is satisfied and that the wasteform will continue to make an appropriate contribution to the overall performance of the waste package, and to the geological disposal system as a whole.

The requirements for wasteforms, and their maintenance for appropriate periods of time, are discussed in more detail in the guidance on the achievement of wasteform performance that supports the WPS, as part of the WPSGD.

6.3 Requirements for waste packages

6.3.1 Activity content

The activity content of the waste package shall be controlled to comply with the radionuclide related assumptions that underpin the safety cases for transport and the GDF operational period.

For waste packages transported as part of a Type B transport package, or as Type B transport packages in their own right the total activity content of the transport package should not exceed 105A2.

The contents of waste packages transported as part of a Type IP transport package, or as Type IP transport packages in their own right, shall be capable of being categorised as low specific activity (LSA) material or as surface contaminated objects (SCO).

The allowable activity content of waste packages may be limited by one or more of a number of radionuclide related parameters which are dealt with separately is this Generic Specification. These comprise:

External dose rate (Section 6.3.3)

Heat output (Section 6.3.4)

17 Hazardous materials include flammable, explosive, pyrophoric, chemo-toxic and oxidising

materials, sealed containers and objects containing stored energy.

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Release of activity in gaseous form (Section 6.3.6)

Criticality safety (Section 6.3.7)

Accident performance (Section 6.3.8)

The IAEA Transport Regulations define limits for the activity contents of different classes of transport packages. In the context of this Generic Specification the relevant limits are:

For Type B transport packages which have not been qualified to satisfy the requirement of an ‘enhanced water immersion test’ a limit of 105A2 is placed on the total activity of their contents; and

The contents of Type IP transport packages must comply with the activity limits and other requirements defined for LSA material and SCOs.

Two categories of LSA material are of relevance to this Generic Specification, LSA-II and LSA-III, for which average specific activity limits of 10-4A2g

-1 and 2x10-3A2g-1 respectively,

are defined. The higher limit only applies to materials where the activity is distributed throughout a solid or is essentially uniformly distributed in a solid compact binding agent such as concrete (i.e. an encapsulated wasteform), and which is relatively insoluble. For non-encapsulated wasteforms the limit for LSA-II material would generally apply. It should also be noted that for LSA-III materials to be carried in Type IP-2 transport packages, transport operations must be conducted under the conditions of exclusive use (see Section 5.1).

SCOs are solid objects which are not intrinsically radioactive but which have radioactive material distributed on their surfaces. This description excludes bulk radioactive material (e.g. uranium metal) and materials such as metals and graphite which have become radioactive as a result of neutron irradiation18.

The total quantity of LSA material or SCOs that can be carried in a Type IP transport package is limited by a maximum dose rate19 defined for the wasteform (i.e. with no benefit claimed for any shielding provided by the waste container). A limit of 100A2 is placed on the quantity of activity that can be associated with solid LSA-II or LSA-III material which is combustible. This limit is reduced to 10A2 if transport is by inland waterway.

6.3.2 Gross mass

The gross mass of the waste package shall be compatible with the transport and GDF handling systems and with the requirement for the waste package to be safely stacked.

The gross mass of a transport package should not exceed 65t.

The maximum gross mass of waste packages must be such that will permit them to be safely and efficiently handled using the systems defined for transport to and emplacement in a GDF. The gross masses of transport packages must also be compatible with the UK transport infrastructure such that no undue limits are placed on the mode of transport that can be used (i.e. by road, rail, sea or inland waterway).

The GDFD [8] currently includes a number of assumptions regarding the safe working loads (SWL) for the GDF lifting and handling equipment. This includes a capability to transfer underground and subsequently handle transport packages with gross masses (including any handling equipment such as lifting frames) of up to 80t.

18 These materials could be carried in Type IP-2 transport packages if they could be shown to be

LSA material by virtue of their average specific activity. 19 Defined in Paragraph 516 as 10mSvh-1 at a distance of 3m from the unshielded wasteform.

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For waste packages which are transport packages in their own right, emplacement is assumed to be by the use of a stacker truck with a SWL of 65t. For waste packages transported within transport containers the lifting and handling equipment used to remove waste packages from the transport container, transfer them to the disposal vaults and stack them therein are assumed to have SWLs of 20t20.

The GTSD [9] currently assumes the use of a four-axle rail wagon (which could be used on a large proportion of the UK rail network) for ILW transport packages, which would limit the gross mass of transport packages to ~64t21. The possibility does exist for the use of eight-axle rail wagons capable of carrying greater loads but these may only be suitable for use on a reduced proportion of the rail network.

The 64t limit is applied directly to the rail transport of waste packages which are transport packages in their own right and to the combined mass of transport containers and their waste package contents. In the latter case, the heaviest of the current range of SWTCs (the SWTC-285) has an empty mass of 52t which places a limit of 12t on the mass of the contents. Heavier waste packages (i.e. up to 20t) could be transported using either of the two less heavily shielded SWTC designs, if they could provide sufficient radiation shielding, but conformance with the requirements for their safe stacking at a GDF (Section 6.1.3) would need to be assured.

For transport by road, the maximum permitted laden mass of an ordinary heavy goods vehicle is 44t which, when an allowance is made for the mass of the vehicle itself, sets a limit of ~30t for the load. Transport packages with gross masses of greater than 30t will require special transport arrangements and there may therefore be operational benefits in maintaining transport package masses below this value, if practicable.

The 30t limit will only permit use of the lightest SWTC (unladen mass ~15t) for the transport of waste packages with masses of up to ~15t. Road transport using heavier SWTCs (i.e. the SWTC-285 or the SWTC-150 which have unladen masses of greater than 30t) would only be possible under special arrangements.

6.3.3 External dose rate

The external dose rate from the waste package shall enable safe handling of the waste package during transport and the GDF operational period, and shall comply with regulatory limits for transport.

The external dose rate of the waste package should be compatible with the dose rate at 1 metre from any external surface of a transport package, under normal conditions of transport, not exceeding 0.1mSvh-1 and the dose rate on its external surface not exceeding 2mSvh-1.

The IAEA Transport Regulations define limits for the external dose rate from transport packages which depend on the operational controls under which transport operations are carried out (i.e. whether under the conditions of ‘exclusive use’ or ‘non-exclusive use’). The more stringent of the limits, those for non-exclusive use, have been adopted as the limiting values for all transport packages in the GTSD [9] and the TSC [11] and are therefore also included in this Generic Specification. These limits require that transport package dose rates are no more than:

2mSvh-1 on the external surface; and

20 This limit applies to complete ‘disposal units’ which may comprise individual waste packages or

groups of waste packages contained in a handling frame or stillage. 21 This value is equal to the maximum permitted gross mass of such a wagon (90t) less the

estimated mass of the unladen wagon (~26t).

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0.1mSvh-1 at 1m from the external surface.

These limits are applied directly to all waste packages transported without additional radiation shielding and indirectly to waste packages carried in transport containers, in which case they apply to the external surfaces of the transport package.

6.3.4 Heat output

The heat generated by the waste package shall be controlled to ensure that:

thermal effects result in no significant deterioration in the performance of the waste package, or of the disposal system as a whole; and

regulatory limits on the surface temperature of transport packages are not exceeded.

The heat output of the waste package at the time of disposal vault closure should not exceed 6 watts per cubic metre of conditioned waste.

Excessive heat generation by waste packages has the potential to result in damage to the waste package itself and could affect the performance of the geological disposal system as a whole, by causing damage to other barriers.

Damage to waste packages, notably as a result of accelerated corrosion of either the waste container or components of the wasteform, could result in an unacceptable degradation of the performance of waste packages. As a means of minimising such damage the DSTS defines a target of 50C for the temperature within ILW disposal vaults during the operational period and uses the same temperature as a guidance value for limiting damage to the GDF barriers in the post-closure period. For disposal vaults that incorporate a cementitious backfill, waste package temperatures of up to 80°C are deemed acceptable in the short-term (i.e. for a period of 5 years) whilst the 50°C target is maintained for the longer-term.

Thermal modelling of a range of possible disposal vault designs has shown that the temperature targets for both the operational and post-backfilling periods would not be threatened by average heat output (i.e. for all of the waste packages in a vault) of 6Wm-3 [32]. This value is therefore used as a screening level for the average heat output of the waste packages that would result from the conditioning of a waste stream. Other work has also shown that individual waste packages with heat outputs of up to ~100Wm-3 would not cause the temperature targets to be exceeded [33].

The IAEA Transport Regulations define qualitative and quantitative controls on the heat generated by the contents of transport packages. These include ensuring that heat generation will not alter the basic physical properties of the transport package or its contents (Paragraph 651).

Paragraphs 652 and 653 specify a maximum temperature of 50°C for any accessible surface of a Type B transport package carried under the conditions of non-exclusive use. Paragraph 562 specifies additional operational controls for the stowage and storage of transport packages that have an average surface heat flux of greater than 15Wm-2. Thermal modelling work has shown that heat generation limits derived from such limits are generally less bounding than those necessary to ensure that the limits and targets specified for the operational period of a GDF [34, 35]. This work has shown that current designs of SWTC could be used to transport waste packages with heat outputs of up to ~400W without exceeding the regulatory temperature or heat flux limits.

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6.3.5 Surface contamination

The non-fixed surface contamination of the waste package shall be as low as reasonably practicable and shall comply with regulatory limits for transport.

For waste packages which are transport packages in their own right the non-fixed surface contamination, when averaged over an area of 300cm2 of any part of the surface of the waste package, shall not exceed:

4.0Bqcm-2 for beta, gamma and low toxicity alpha emitters22; and

0.4Bqcm-2 for all other alpha emitters.

For waste packages transported inside transport containers the non-fixed surface contamination, when averaged over an area of 300cm2 of any part of the surface of the waste package, should not exceed:

4.0Bqcm-2 for beta, gamma and low toxicity alpha emitters; and

0.4Bqcm-2 for all other alpha emitters.

Limits on the non-fixed23 surface contamination of waste packages are specified to ensure that:

Regulatory limits are achieved for waste packages which are transported without additional protection;

Contamination of transport and GDF systems can be maintained at appropriate levels; and

Routine doses to workers and the members of the public will be ALARP and in accordance with good industry practice.

The limits specified are those defined in Paragraph 507 of the IAEA Transport Regulations for the non-fixed surface contamination of transport packages and therefore are applied directly to all waste packages transported without additional protection.

The same limits are also used as targets for the non-fixed contamination of all waste packages on the basis that they represent realistic and achievable levels and will reduce any potential requirement for the decontamination of the internal surfaces of transport containers and the areas of a GDF where ‘bare’ waste package are handled.

The surface contamination limits only apply to ‘non-fixed’ contamination on the basis that such material could become detached from the waste package during routine operations and inhaled or ingested by humans. Contamination deemed as being ‘fixed’ cannot be as readily removed and therefore cannot cause dose by such mechanisms. It can however contribute to the external radiation from the waste packages and is covered by the limits defined in Section 6.3.2. It should be noted however that ‘fixed’ contamination can become ‘non-fixed’ as a result of the effects of waste package ageing, weather or handling and that the level of non-fixed contamination could increase with time.

22 Low toxicity alpha emitters are defined by the IAEA Transport Regulations (Paragraph 227) as

‘…natural uranium; depleted uranium; natural thorium; uranium-235 or uranium-238; thorium-232; thorium 228 and thorium-230 when contained in ores or physical and chemical concentrates; or alpha emitters with a half-life of less than 10 days’.

23 Non-fixed contamination is defined by the IAEA Transport Regulations (Paragraph 215) as ‘…contamination that can be removed from a surface during routine conditions of transport’.

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6.3.6 Gas generation

The generation of bulk, radioactive and toxic gases by the waste package shall comply with the requirements for safe transport and disposal.

The release of radionuclides in gaseous form from the waste package shall comply with the assumptions that underpin the safety cases for transport and the GDF operational period.

Gases generated by waste packages transported as part of a Type B transport package, or as Type B transport packages in their own right, shall not:

cause the internal pressure of the transport package to exceed a gauge pressure of 700kPa under normal conditions of transport; or

result in the release of radionuclides, in gaseous or particulate form, from the transport package under normal conditions of transport exceeding 10-6A2 per hour.

Gases generated by waste packages transported as part of a Type IP transport package, or as Type IP transport packages in their own right, should not:

cause the internal pressure of the transport package to exceed a gauge pressure of 700kPa under normal conditions of transport; or

result in the release of radionuclides, in gaseous or particulate form, from the transport package under normal conditions of transport exceeding 10-6A2 per hour.

The release of activity, in gaseous or particulate form, from the waste package during the GDF operational period should not exceed:

8kBq/hour per cubic metre of conditioned waste for hydrogen-3;

180Bq/hour per cubic metre of conditioned waste for carbon-14; or

150Bq/hour per cubic metre of conditioned waste for radon-222.

The physical, chemical, biological and radiological properties of the wastes covered by this Generic Specification are such that the potential exists for the generation of a wide range of gases by a number of different processes including radioactive decay, radiolysis and corrosion and other processes that result in the release of entrained radioactive gases from wasteforms.

The most significant ‘bulk’ gases generated by wastes are H2, CO2 and CH4 which can give rise to a range of potential effects that may have an influence on all periods of the long-term management of waste packages. In the early stages of their management these potential effects include pressurisation and damage to waste containers and wasteforms, pressurisation of sealed transport containers and the release of radioactive, toxic and flammable gases from waste packages and transport packages. After closure of a GDF gas generation has the potential to cause pressurisation and damage to the EBS and host rocks leading to the potential modification of groundwater flow patterns, the rate of re-saturation of disposal vaults and/or mineralogical changes of backfill material.

The IAEA Transport Regulations place limits on the generation of gas by waste packages by limiting the internal pressure of transport packages to a maximum normal operating pressure24 (MNOP) and the loss of the radioactive contents (which will include both radioactive gases and activity in particulate form which may be entrained in non-radioactive

24 Paragraph 661 requires that the MNOP for a Type B transport package shall not exceed 800kPa

absolute pressure. The same maximum value of MNOP is assumed to apply to waste packages transported as Type IP transport packages.

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gases) under normal conditions of transport to 10-6A2 per hour. These limits apply to Type B transport packages and whilst not applicable to Type IP-2 transport packages they are used in this Generic Specification as performance targets for waste packages transported as such. Specifically the activity release rate of 10-6A2 per hour is used to quantify the requirement to prevent the loss or dispersal of the radioactive contents from Type IP transport packages under normal conditions of transport.

Sealed waste packages which are transport packages in their own right will be required to have a MNOP defined and justified by the relevant design authority. All sealed waste packages will require the definition of a ‘design pressure constraint’ to ensure that the consequences of internal gas generation are managed safely during the GDF operational period.

During the operational period of a GDF the ventilation system will prevent unsafe accumulations of toxic, asphyxiating, radioactive, flammable or explosive gases within the disposal vaults and associated facilities by managing them to safe concentrations and discharging them to the atmosphere.

In the post-closure period the migration of gases from the disposal vaults is one of the main potential pathways by which radionuclides, and other hazardous materials, might be released to the accessible environment. Gases produced by waste packages in this period could thus have a significant effect on post-closure safety, if the potential for their generation is not managed appropriately at the packaging stage.

The release of activity in gaseous form from waste packages has the potential to cause on- and off-site dose during both the GDF operational and post-closure periods. The ESC [13] identifies hydrogen-3, carbon-14 and radon-222 as the three most significant radionuclides that could be released from waste packages in gaseous form in such a manner that could lead to off-site dose. The generic Operational Environmental Safety Assessment [36] uses a value of 0.01mSv/year (derived from the 2009 Statutory Guidance to the Environment Agency [37]) as a target for the maximum dose to the most exposed group of members of the public due to routine discharges from a GDF. This value is used to define screening levels for the release of gaseous radionuclides from waste packages on the basis that if these levels were exceeded by the entire ILW inventory (i.e. ~360,000m3) the 0.01mSv/year target would be exceeded.

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6.3.7 Criticality safety

The presence of fissile material, neutron moderators and reflectors in the waste package shall be controlled to ensure that:

criticality during transport is prevented;

the risk of criticality during the GDF operational period is tolerable and as low as reasonably practicable; and

in the GDF post-closure period both the likelihood and the consequences of a criticality are low.

The total quantity of fissile material in the waste package should not exceed 47g.

The quantities of fissile material, neutron moderators and reflectors in the waste package shall be controlled to ensure that the transport package satisfies the criticality safety requirements of the IAEA Transport Regulations.

For waste packages transported as part of a Type IP transport package, or as a Type IP transport package in their own right, the quantities of fissile material, neutron moderators and reflectors in the waste package should be controlled to ensure that the transport package can be excepted from the requirements of the IAEA Transport Regulations for packages containing fissile material.

The UK ILW inventory includes a large number of waste streams which contain significant quantities of fissile material, mainly in the form of surface contamination and discrete pieces of irradiated uranium fuel. The quantities of such materials in waste packages must be controlled to ensure that they do not represent an unacceptable criticality safety hazard at any stage during their long-term management, including during the GDF post-closure period.

The passive criticality safety of waste packages is generally achieved by controlling the quantities of fissile material and of neutron moderating and reflecting materials25 they contain, such that criticality cannot occur under all credible conditions during their long-term management. This relies on determining the limiting quantities of fissile materials that will satisfy the criticality safety requirements for transport and the operational and post-closure periods of a GDF for the waste package, and ensuring that this quantity is not exceeded during waste package manufacture.

Work to determine waste package fissile materials limits which take into account all of the three requirements listed above has led to the definition of a ‘General Screening Level’ (GSL) of 47g26 for all existing designs of unshielded waste package27 [38]. The GSL applies to all wastes covered by this Generic Specification, irrespective of the nature of the fissile material, and places relatively minor constraints on the quantities of neutron reflectors and moderators that could be present in a waste package. The GSL represents a fissile material limit which can be applied to most waste packages without the need for further justification (e.g. by way of waste characterisation).

For wastes which are more characterised, and/or where more restrictive constraints can be placed on the quantities of neutron reflectors and moderators, significantly larger quantities of fissile material can be safely accommodated in waste packages. In general, this will require a better understanding of the nature of the fissile material (e.g. the proportions of

25 In general the three materials of interest are graphite, deuterated compounds and beryllium. 26 This limit being the mass of plutonium-239, or the total mass of all fissile nuclides which would

produce the equivalent reactivity of 47g of plutonium-239 with optimal shape and neutron moderation and reflection.

27 These comprise the 500 litre drum, 3 cubic metre box and 3 cubic metre drum waste packages.

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different fissile nuclides present) and, in particular, will need to claim benefit for the presence of neutron poisons such as uranium-238.

With specific regard to transport, the IAEA Transport Regulations define requirements for transport packages containing fissile material (Paragraphs 671 to 683) which can be used to define the permitted quantities of fissile material that can be carried. The methodology underpinning the derivation of the GSL reflects these requirements although, as above, significantly greater quantities of fissile material can be transported if appropriate characterisation of the waste can be demonstrated.

Transport packages containing very small quantities of fissile material can be excepted from the requirements defined in Paragraphs 671 to 683. Paragraph 417 defines a number of mass limits for certain categories of material containing fissile radionuclides that can be so excepted. Additionally, Paragraph 222 excludes from the definition of fissile material natural uranium or depleted uranium which is unirradiated, or which has been irradiated in thermal reactors only.

During the development of proposals to package wastes containing fissile material waste packagers will be required to demonstrate that the requirement for the criticality safety of the proposed waste packages will be satisfied in practice (i.e. during packaging). To this end, and as part of a submission for the disposability assessment of the proposed waste packages, the waste packager will need to produce Criticality Compliance Assurance Documentation (CCAD). The CCAD will need to consider the quantity and form of the fissile materials in a waste stream, identify and justify a suitable safe fissile mass (SFM) and define the procedural controls will be put in place to ensure that the SFM will not exceeded during the packaging of the waste. Guidance is available on the requirements and preferred format of CCAD [39].

6.3.8 Accident performance Under all credible accident scenarios the release of radionuclides and other hazardous materials from the waste package shall be low and predictable.

The waste package should exhibit progressive release behaviour within the range of all credible accident scenarios.

The impact and fire accident performance of the waste package shall comply with the assumptions that underpin the safety cases for transport and the GDF operational period.

The accident performance of the waste package shall ensure that, in the event of any credible accident during the GDF operational period, the on- and off-site doses resulting from the release of radionuclides from the waste package shall be as low as reasonably practicable and should be consistent with meeting the relevant Basic Safety Levels.

Waste packages may be subject to a range of accident conditions during their long-term management, up until the end of the GDF operational period. Such accidents, which notably include impacts and fires, are a potential mechanism for the release of radionuclides from waste packages into the environment. Waste packages must be capable of complying with the assumptions that are made in the safety cases for transport and the GDF operational period regarding their performance in response to specified accident scenarios. Additionally the radiation doses to workers and members of the public resulting from the release of radionuclide from waste packages as a consequence of accidents must be ALARP and less than the relevant regulatory limits.

That magnitude of the mechanical, thermal and other challenges that could result from accidents at a GDF will vary depending on the design of the facilities, and the manner in which the facility is operated. The response of waste packages to the full range of such challenges should be ‘progressive’ in that there should be no significant ‘cliff-edge’ effects

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where a small increase in the magnitude of the challenge would produce a large deterioration in waste package performance.

The IAEA Transport Regulations define ‘accident conditions of transport’ as a basis for demonstrating the ability of transport packages to withstand impact and fire accidents during transport without causing excessive exposure of transport workers or members of the public to radionuclides released during such accidents. The treatment of the different types of transport package varies significantly with regard to accident performance. For Type B transport packages, safety under accident conditions is conferred by defining requirements for the performance of the containment and for the physical protection provided by the transport package itself. By contrast, for Type IP transport packages, safety is conferred by applying restrictions on the form and specific activity of their radionuclide contents.

Type B transport packages are required to be capable of being subject to the cumulative effects of defined mechanical and thermal challenges28 following which the accumulated loss of activity from the transport package in a period of one week must not exceed 1A2

29. Waste packages qualifying as Type B transport packages in their own right would be required to satisfy this requirement. For waste packages transported within transport containers, the transport package as a whole must satisfy the requirement and significant benefit can be claimed for the protection provided.

The IAEA Transport Regulations do not specify accident conditions of transport for Type IP transport packages, although the normal conditions of transport for such packages require that they should be capable of withstanding a ‘free drop test30’, without suffering any loss or dispersal31 of the radioactive contents or significant loss of shielding integrity32.

During the GDF operational period the potential exists for a range of accidents which could result in damage to waste packages, the release of radionuclides and radiation dose to workers on-site and/or members of the public off-site. As well as requiring that the doses resulting from accidents in which radionuclides are released are ALARP the HSE Safety Assessment Principles (SAPs) [40] define Basic Safety Objectives (BSOs) for the cumulative annual on- and off-site doses due to accidents on nuclear sites. The SAPs also define Basic Safety Levels (BSLs) as targets for the maximum on- and off-site dose that could result from release of radionuclides as a result of design basis accidents, on the basis of the expected frequency of the initiating event that would result in such an accident. The highest of these BSLs, for accidents with an expected initiating event frequency of less than 10-4 per annum, are therefore used as upper bounds for the dose consequences of GDF accidents. However, for accidents for which a higher fault frequency cannot be discounted, the lower BSLs will apply.

28 Comprising a 9m drop on to a flat unyielding surface, a 1m drop on to an aggressive feature, a

dynamic crush test and exposure to a fire with a flame temperature of 800C for 30 minutes. 29 Or not more than 10A2 for krypton-85. 30 The drop height for such a test varies from 1.2m for transport packages with a mass of less than

5t, to 0.3m for packages with masses of greater than 15t. 31 We would generally assume this requirement to be satisfied if the loss of activity from the waste

packages was less than 10-6 A2 per hour following the impact. 32 Such as would result in more than a 20% increase in the radiation level at any external surface of

the transport package.

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6.4 Requirements for the manufacture and storage of waste packages

Adequate controls shall be established and applied to ensure that manufactured waste packages have the properties and performance required of them.

Adequate controls shall be applied during any period of interim storage to ensure that waste packages retain their required properties and performance for the duration of such a period.

6.4.1 Quality management

Adequate management arrangements shall be applied to all aspects of the packaging of radioactive waste, and the storage of waste packages, that affect product quality. These arrangements shall be agreed with RWMD prior to the start of the activities to which they relate.

All activities relevant to licensing of a GDF will be conducted in accordance with appropriate quality management arrangements. The objective in establishing and operating a quality management system is to provide an integral framework of procedures which will ensure that the work is adequately controlled, documented and recorded. It is the responsibility of the SLC to develop, operate and maintain appropriate quality management arrangements for all aspects of the retrieval and packaging of waste and the interim storage of waste packages. As a minimum all quality management systems shall comply with BS EN ISO9001 [41] and should comply with RWMD specification [42] and its supporting guidance [43].

6.4.2 Waste package data and information recording

Information shall be recorded for each waste package covering all relevant details of its manufacture and interim storage. This information shall be sufficient to enable assessment of the characteristics and performance of the waste package against the requirements of all stages of long-term management.

Information shall be recorded regarding the quantity of all radionuclides of relevance to the safe transport and disposal of the waste package.

Compliance with the various regulations that apply to the transport and disposal of radioactive waste will rely in part on the existence of appropriate records regarding all relevant aspects of waste package manufacture, notably information regarding their radionuclide inventory. The recording of such data and information should be carried out in accordance with the RWMD specification [44].

Not all of the radionuclides that could be present in the wastes to which this Generic Specification applies will have relevance to the safety of long-term management of those wastes. A methodology has been developed for the identification of those radionuclides which have relevance to the safety of one or more of the three main stages of geological disposal of ILW (i.e. transport, and the GDF operational and post-closure periods) [45]. Waste producers will be required to identify which of those radionuclides will be present in the waste to be packaged at levels which will require their quantities to be determined and recorded during waste package manufacture.

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6.4.3 Requirements for waste packages containing nuclear material

The management of waste packages containing nuclear material shall comply with all relevant international safeguards obligations and security requirements for their transport and disposal.

Many of the wastes to which this Generic Specification applies will contain radionuclides that are described as ‘nuclear materials’ (NM)33. All operations involving NM are subject to regulatory control and from the viewpoint of the long-term management of wastes containing NM this will include retrieval from storage, packaging, and the interim storage, transport and disposal of waste packages. Waste producers will be required to address all of the requirements for nuclear materials accountancy and the security of NM up to the point of their receipt at a GDF site and also will need to ensure that appropriate records are available to permit RWMD to assume responsibility for the NM in waste packages from that point.

International safeguards The safeguards status of any nuclear material contained within the waste package shall be ascertained and recorded.

Packaged wastes that contain NM derived from the UK civil nuclear programme may be subject to controls known as 'safeguards' allowing independent verification by international nuclear safeguards inspectorates (i.e. Euratom and IAEA) to confirm that nuclear material has not been diverted from peaceful use.

In order that the contents of waste packages that contain safeguarded materials can be fully verified, SLCs will be required to provide sufficient information on the quantity, nature and status of all safeguarded material that will be incorporated into proposed waste packages.

IAEA has published a comprehensive overview of the techniques and equipment underlying the implementation of IAEA safeguards, including those used for nuclear material accountancy, containment and surveillance measures, environmental sampling, and data security [46]. The HSE has also produced guidance on good practice for nuclear material accountancy and safeguards on nuclear licensed sites [47].

Physical protection nuclear security The quantity of nuclear material contained within the waste package shall be controlled such that the waste package can be transported subject to standards of physical protection no higher than those defined for the transport system.

The Nuclear Industries Security Regulations 2003 (NISR) lay down the requirements for security of nuclear premises, security of transport of nuclear material and security of sensitive nuclear information. The NISR are administered and enforced by ONR acting on behalf of the Secretary of State for Energy and Climate Change.

IAEA guidance on the physical protection of NM [48], which accords with NISR in this regard, specifies mass limits for the quantities of NM that can be transported with three ‘categories’ of physical protection (Categories I to III, Category I being the most restrictive). It is currently our intention that all waste packages destined for emplacement in a GDF will be permitted to be transported with protection to standards no higher than those defined by Category II.

33 NM are defined by the Office for Civil Nuclear Security (OCNS) as plutonium, uranium,

neptunium, americium, and thorium.

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7 Summary

This Generic Specification defines requirements for all waste packages containing waste generally characterised as being of low heat generation, and that will be the subject of geological disposal. The requirements derive directly from those defined for waste packages by the generic DSS and apply the high-level requirements defined by the GWPS to the waste packages containing such wastes.

The form of this Generic Specification is to define bounding requirements for waste packages such that it can be used as the basis for the development of proposals to package low heat generating waste using a wide range of waste containers and waste conditioning processes. To assist in the development of such proposals users are referred to the WPSGD, specifically to the guidance which explains the basis for the derivation of the packaging requirements and the manner in which they can be achieved for practical packaging projects.

This Generic Specification acts as the basis for the definition of the WPS which define the requirements for the waste packages containing low heat generating waste that would result from the use of standardised designs of waste container.

This Generic Specification also provides a baseline for the conduct of Conceptual stage disposability assessments of proposals to package low heat generating waste for geological disposal.

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31 Serco, Implications of RWMD 500 year waste container integrity target compared with 150 years for container design and cost, SERCO/005084/001, 2011.

32 Electrowatt-Ekono, Thermal Analysis of 7 by 7 Stillage Arrays in ILW Vaults, EWE-17314/002 Revision 4, 2001.

33 Serco Assurance, 3-D Thermal Modelling of Waste Packages in Backfilled Vaults, SERCO/TAS/2584/W1, 2008.

34 Serco Assurance, Thermal Analysis of the SWTC-70 Concept Design, SA/PSS/14781/W2, 2004.

35 Nirex, Radiogenic Heating Limits for 4 metre ILW Boxes, T/REP/20641,1998.

36 NDA, Geological Disposal: Generic Operational Environmental Safety Assessment, NDA/RWMD/029, 2010.

37 Secretary of State for Energy and Climate Change, Secretary of State for Health in relation to England, and the Welsh Ministers, Statutory Guidance to the Environment Agency concerning the regulation of radioactive discharges into the environment, 2009.

38 T.W. Hicks, The General Criticality Safety Assessment, Galson Sciences, 0914-1 Version 1.1, 2009.

39 NDA, Guidance on the preparation of Criticality Compliance Assurance Documentation for waste packaging proposals, WPS/625/02, 2008.

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40 HSE, Nuclear Installations Inspectorate: Safety Assessment Principles for Nuclear Facilities, 2006 Edition, Revision 1, 2006.

41 British Standards Institution, Quality Management Systems - Requirements, BS EN ISO9001.

42 NDA, Waste Package Quality Management Specification, WPS/200/02, 2008.

43 NDA, Waste Package Quality Management Specification: Guidance Material, WPS/210/02, 2008.

44 NDA, Waste Package Data and Information Recording Specification, WPS/400/02, 2008.

45 Nirex, The Identification of Radionuclides Relevant to Long-term Waste Management in the UK, Nirex Report N/105, 2004.

46 IAEA, Safeguards Techniques and Equipment: 2011 Edition, International Nuclear Verification Series No. 1 (Rev. 2), 2011.

47 HSE, Guidance on International Safeguards and Nuclear Material Accountancy at Nuclear Sites in the UK, 2010 edition, Revision 1.

48 IAEA, Guidance and Consideration of the Implementation on INFCIRC/225/Rev 4, The Physical Protection of Nuclear Materials and Nuclear Facilities, IAEA-TECDOC-967 (Rev. 1), 2000.

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Appendix A Glossary of terms used in this document

activity

The number of atoms of a radioactive substance which decay by nuclear disintegration each second. The SI unit of activity is the becquerel (Bq) equal to one radioactive decay per second.

The IAEA Transport Regulations define a unit of activity, the A2, as a means of standardising the dose consequences of different radionuclides on the basis of the different possible exposure pathways that could occur following the release of radionuclides from a transport package. A2 values (in TBq) for a wide range of radionuclides are listed in Table 2 of the IAEA Transport Regulations [19].

alpha activity

Alpha activity takes the form of particles (helium nuclei) ejected from a decaying (radioactive) atom. Alpha particles cause ionisation in biological tissue which may lead to damage. The particles have a very short range in air (typically about 5cm) and alpha particles present in materials that are outside of the body are prevented from doing biological damage by the superficial dead skin cells, but become significant if inhaled or swallowed.

backfill

A material used to fill voids in a GDF. Three types of backfill are recognised:

local backfill, which is emplaced to fill the free space between and around waste packages;

peripheral backfill, which is emplaced in disposal modules between waste and local backfill, and the near-field rock or access ways; and

mass backfill, which is the bulk material used to backfill the excavated volume apart from the disposal areas.

backfilling

The refilling of the excavated portions of a disposal facility after emplacement of the waste.

barrier

A physical or chemical means of preventing or inhibiting the movement of radionuclides.

beta activity

Beta activity takes the form of particles (electrons) emitted during radioactive decay from the nucleus of an atom. Beta particles cause ionisation in biological tissue which may lead to damage. Most beta particles can pass through the skin and penetrate the body, but a few millimetres of light materials, such as aluminium, will generally shield against them.

buffer

An engineered barrier that protects the waste package and limits the migration of radionuclides following their release from a waste package.

canister

A term used in specific concepts to describe the waste container into which a wasteform is placed.

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conditioning

Treatment of a radioactive waste material to create, or assist in the creation of, a wasteform that has passive safety

container

The vessel into which a wasteform is placed to form a waste package suitable for handling, transport, storage and disposal.

containment

The engineered barriers, including the waste form and packaging, shall be so designed, and a host geological formation shall so be selected, as to provide containment of the waste during the period when waste produces heat energy in amounts that could adversely affect the containment, and when radioactive decay has not yet significantly reduced the hazard posed by the waste

criticality

A state in which a quantity of fissile material can maintain a self-sustaining neutron chain reaction. Criticality requires that a sufficiently large quantity of fissile material (a critical mass) be assembled into a geometry that can sustain a chain reaction; unless both of these requirements are met, no chain reaction can take place and the system is said to be sub-critical.

criticality safety

A methodology used to define the conditions required to ensure the continued sub-criticality of waste containing fissile material.

disposability

The ability of a waste package to satisfy the defined requirement for disposal.

disposability assessment

The process by which the disposability of proposed waste packages is assessed. The outcome of a disposability assessment may be a Letter of Compliance endorsing the disposability of the proposed waste packages.

disposal

In the context of solid waste, disposal is the emplacement of waste in a suitable facility without intent to retrieve it at a later date; retrieval may be possible but, if intended, the appropriate term is storage.

disposal canister

A term used to describe the assembly of certain waste types (e.g. HLW, spent fuel, plutonium, HEU) within a metal container, as prepared for disposal.

disposal facility (for solid radioactive waste)

An engineered facility for the disposal of solid radioactive wastes.

disposal system

All the aspects of the waste, the disposal facility and its surroundings that affect the radiological impact.

disposal unit

A waste package, or group of waste packages, which is handled as a single unit for the purposes of transport and/or disposal.

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disposal vault

Underground opening where ILW or LLW waste packages are emplaced.

dose

A measure of the energy deposited by radiation in a target.

dose rate

The effective dose equivalent per unit time. Typical units of effective dose are sievert/hour (Svh-1), millisieverts/hour (mSvh-1) and sievert/year (Svy-1).

emplacement (of waste in a disposal facility)

The placement of a waste package in a designated location for disposal, with no intent to reposition or retrieve it subsequently.

Environment Agency (EA)

The environmental regulator for England and Wales. The Agency’s role is the enforcement of specified laws and regulations aimed at protecting the environment, in the context of sustainable development, predominantly by authorising and controlling radioactive discharges and waste disposal to air, water (surface water, groundwater) and land. The Environment Agency also regulates nuclear sites under the Environmental Permitting Regulations and issues consents for non-radioactive discharges.

environmental safety case

The collection of arguments, provided by the developer or operator of a disposal facility, that seeks to demonstrate that the required standard of environmental safety is achieved.

fissile material

Fissile material is that which undergoes fission under neutron irradiation. For regulatory purposes material containing any of the following nuclides is considered to be ‘fissile’: uranium-233, urainium-235, plutonium-239 and plutonium-241.

gamma activity

An electromagnetic radiation similar in some respects to visible light, but with higher energy. Gamma rays cause ionisations in biological tissue which may lead to damage. Gamma rays are very penetrating and are attenuated only by shields of dense metal or concrete, perhaps some metres thick, depending on their energy. Their emission during radioactive decay is usually accompanied by particle emission (beta or alpha activity).

geological disposal

A long term management option involving the emplacement of radioactive waste in an engineered underground geological disposal facility or repository, where the geology (rock structure) provides a barrier against the escape of radioactivity and there is no intention to retrieve the waste once the facility is closed.

geological disposal facility (GDF)

An engineered underground facility for the disposal of solid radioactive wastes.

half-life

The time taken for the activity of a given amount of a radioactive substance to decay to half of its initial value. Each radionuclide has a unique half-life.

hazardous materials

Materials that can endanger human health if improperly handled. As defined by the Control of Substances Hazardous to Health Regulations, 2002.

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Health and Safety Executive (HSE)

The HSE is a statutory body whose role is the enforcement of work-related health and safety law. HSE is formally the licensing authority for nuclear installations in Great Britain, although the licensing function is administered on HSE's behalf by its executive agency the Office for Nuclear Regulation (ONR).

higher activity radioactive waste

Generally used to include the following categories of radioactive waste: low level waste not suitable for near surface disposal, intermediate level waste and high level waste.

high level waste (HLW)

Radioactive wastes in which the temperature may rise significantly as a result of their radioactivity, so this factor has to be taken into account in the design of storage or disposal facilities.

immobilisation

A process by which the potential for the migration or dispersion of the radioactivity present in a material is reduced. This is often achieved by converting the material to a monolithic form that confers passive safety to the material.

Industrial Package (Type-IP)

A category of transport package, defined by the IAEA Transport Regulations for the transport of radioactive materials with low specific activities.

intermediate level waste (ILW)

Radioactive wastes exceeding the upper activity boundaries for LLW but which do not need heat to be taken into account in the design of storage or disposal facilities.

International Atomic Energy Agency (IAEA)

The IAEA is the world’s centre of cooperation in the nuclear field. It was set up as the world’s "Atoms for Peace" organization in 1957 within the United Nations family. The Agency works with its Member States and multiple partners worldwide to promote safe, secure and peaceful nuclear technologies.

Letter of Compliance (LoC)

A document, prepared by RWMD, that indicates to a waste packager that a proposed approach to the packaging of waste would result in waste packages that are compliant with the requirements defined by relevant packaging specifications, and the safety assessments for transport to and disposal in a GDF, and are therefore deemed ‘disposable’.

low level waste (LLW)

Radioactive waste having a radioactive content not exceeding 4 gigabecquerels per tonne (GBq/t) of alpha or 12 GBq/t of beta/gamma activity.

Low Level Waste Repository (LLWR)

The UK national facility for the near surface disposal of solid LLW, located near to the village of Drigg in Cumbria.

low specific activity (LSA) material

A material classification defined by the IAEA Transport Regulations as ‘Radioactive material which by its nature has a limited specific activity (i.e. activity per unit mass of material), or radioactive material for which limits of estimated average specific activity apply.’

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Managing Radioactive Waste Safely (MRWS)

A phrase covering the whole process of public consultation, work by CoRWM, and subsequent actions by Government, to identify and implement the option, or combination of options, for the long term management of the UK’s higher activity radioactive waste.

Nirex (United Kingdom Nirex Limited)

An organisation previously owned jointly by Department for the Environment, Food and Rural Affairs and the Department for Trade and Industry. Its objectives were, in support of Government policy, to develop and advise on safe, environmentally sound and publicly acceptable options for the long-term management of radioactive materials in the United Kingdom. The Government’s response to Committee on Radioactive Waste Management in October 2006 initiated the incorporation of Nirex functions into the NDA, a process which was completed in March 2007.

Nuclear Decommissioning Authority (NDA)

The NDA is the implementing organisation, responsible for planning and delivering the GDF. The NDA was set up on 1 April 2005, under the Energy Act 2004. It is a non-departmental public body with designated responsibility for managing the liabilities at specific sites. These sites are operated under contract by site licensee companies (initially British Nuclear Group Sellafield Limited, Magnox Electric Limited, Springfields Fuels Limited and UK Atomic Energy Authority). The NDA has a statutory requirement under the Energy Act 2004, to publish and consult on its Strategy and Annual Plans, which have to be agreed by the Secretary of State (currently the Secretary of State for Trade and Industry) and Scottish Ministers.

Nuclear Installations Act 1965 (NIA65)

UK legislation which provides for the operation and regulation of nuclear installations within the UK.

Nuclear Installations Inspectorate (NII)

See HSE

nuclear material

Fissile material or material that can be used to produce fissile material (i.e. source material). This includes all isotopes of uranium, plutonium and thorium, together with certain isotopes of neptunium and americium.

Office for Civil Nuclear Security (OCNS)

The ONR’s Office for Civil Nuclear Security (OCNS) is the security regulator for the UK's civil nuclear industry. It is responsible for approving security arrangements within the industry and enforcing compliance. OCNS conducts its regulatory activities on behalf of the Secretary of State for Energy and Climate Change (DECC) under the authority of the Nuclear Industries Security Regulations 2003 (NISR 03). OCNS also undertakes vetting of nuclear industry personnel with access to sensitive nuclear material or information. OCNS works in close conjunction with nuclear security policy officials in DECC and with other government departments and agencies, and with overseas counterparts.

Office for Nuclear Regulation (ONR)

The HSE’s executive agency ONR is responsible for regulating the nuclear, radiological and industrial safety of nuclear installations and the transport of radioactive materials in Great Britain under the Nuclear Installations Act 1965 (NIA 65) and the Carriage of Dangerous Good Regulations.

The Government intends to bring forward legislation to establish ONR as a new independent statutory body outside of the HSE to regulate the nuclear power industry,

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formally responsible in law for delivering regulatory functions. The creation of the ONR as a statutory body will consolidate the regulation of civil nuclear and radioactive transport safety and security regulation through one organisation. Pending the legislation, and in the interim, the HSE has established the ONR as a non-statutory body. The Government will review the functions and processes of the interim body in order to inform its planned legislation.

operational period (of a disposal facility)

The period during which a disposal facility is used for its intended purpose, up until closure.

package

See waste package, transport package.

passive safety

Not placing reliance on active safety systems and human intervention to ensure safety.

plutonium (Pu)

A radioactive element occurring in very small quantities in uranium ores but mainly produced artificially, including for use in nuclear fuel, by neutron bombardment of uranium.

post-closure period (of a disposal facility)

The period following sealing and closure of a facility and the removal of active institutional controls.

quality management system (QMS)

A quality management system is the overall system by which an organisation determines, implements and ensures quality.

radioactive decay

The process by which radioactive material loses activity, e.g. alpha activity naturally. The rate at which atoms disintegrate is measured in becquerels.

radioactive material

Material designated in national law or by a regulatory body as being subject to regulatory control because of its radioactivity.

radioactive waste

Any material contaminated by or incorporating radioactivity above certain thresholds defined in legislation, and for which no further use is envisaged, is known as radioactive waste.

Radioactive Waste Management Directorate (RWMD)

The NDA Directorate established to design and build an effective delivery organisation to implement a safe, sustainable, publicly acceptable geological disposal programme. It is envisaged that this directorate will become a wholly owned subsidiary company of the NDA. Ultimately, it will evolve under the NDA into the organisation responsible for the delivery of the GDF. Ownership of this organisation can then be opened up to competition, in due course, in line with other NDA sites.

radioactivity

Atoms undergoing spontaneous random disintegration, usually accompanied by the emission of radiation.

radionuclide

A radioactive form of an element, for example carbon-14 or caesium-137.

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retrievability

A feature of the design of a GDF that enables the waste to be withdrawn, even after the disposal vaults have been backfilled

safeguards

Measures used to verify that nation states comply with their international obligations not to use nuclear materials (plutonium, uranium and thorium) for nuclear explosives purposes. Global recognition of the need for such verification is reflected in the requirements of the Treaty on the Non-Proliferation of Nuclear Weapons (NPT) for the application of safeguards by the International Atomic Energy Agency. Also, the Treaty Establishing the European Atomic Energy Community (the Euratom Treaty) includes requirements for the application of safeguards by the European Community.

safety case

A ‘safety case’ is the written documentation demonstrating that risks associated with a site, a plant, part of a plant or a plant modification are as low as reasonably practicable and that the relevant standards have been met. Safety cases for licensable activities at nuclear sites are required as license conditions under NIA65.

safety function

A specific purpose that must be accomplished for safety.

Scottish Environment Protection Agency (SEPA)

The environmental regulator for Scotland. SEPA’s role is the enforcement of specified laws and regulations aimed at protecting the environment, in the context of sustainable development, predominantly by authorising and controlling radioactive discharges and waste disposal to air, water (surface water, groundwater) and land. SEPA also regulates nuclear sites under the Pollution Prevention and Control Regulations and issues consents for non-radioactive discharges.

shielded waste package

A shielded waste package is one that either has in-built shielding or contains low activity materials, and thus may be handled by conventional techniques.

shielding

Shielding is the protective use of materials to reduce the dose rate outside of the shielding material. The amount of shielding required to ensure that the dose rate is as low as reasonably practicable (ALARP) will therefore depend on the type of radiation, the activity of the source, and on the dose rate that is acceptable outside the shielding material.

spent nuclear fuel

Nuclear fuel removed from a reactor following irradiation that is no longer usable in its present form because of depletion of fissile material, poison build-up or radiation damage.

stack (of waste packages)

A stack of waste packages placed vertically one on top of each other.

surface contaminated object (SCO)

A solid object which is not itself radioactive but which has radioactive material distributed on its surfaces.

transport container

A reusable container into which waste packages are placed for transport, the whole assembly then being referred to as a transport package.

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transport package

The complete assembly of the radioactive material and its outer packaging, as presented for transport.

Transport Regulations

The IAEA Regulations for the Safe Transport of Radioactive Material and/or those regulations as transposed into an EU Directive, and in turn into regulations that apply within the UK. The generic term ‘Transport Regulations’ can refer to any or all of these, since the essential wording is identical in all cases.

transport system

The transport system covers the transport modes, infrastructure, design and operations. It can be divided in two main areas– the transport of construction materials, spoil and personnel associated with building a GDF and the more specialised transport of the radioactive waste to a GDF by inland waterway, sea, rail and/or road.

unshielded waste package

A waste package which, owing either to radiation levels or containment requirements, requires remote handling and must be transported in a reusable transport container.

uranium (U)

A heavy, naturally occurring and weakly radioactive element, commercially extracted from uranium ores. By nuclear fission (the nucleus splitting into two or more nuclei and releasing energy) it is used as a fuel in nuclear reactors to generate heat.

Uranium is often categorised by way of the proportion of the radionuclide uranium-235 it contains (see natural uranium, depleted uranium, low enriched uranium and highly enriched uranium)

waste acceptance criteria (WAC)

Quantitative and/or qualitative criteria, specified by the operator of a disposal facility and approved by the regulator, for solid radioactive waste to be accepted for disposal.

Quantitative or qualitative criteria specified by the regulatory body, or specified by an operator and approved by the regulatory body, for radioactive waste to be accepted by the operator of a repository for disposal, or by the operator of a storage facility for storage.

waste container

Any vessel used to contain a wasteform for disposal.

wasteform

The waste in the physical and chemical form in which it will be disposed of, including any conditioning media and container furniture (i.e. in-drum mixing devices, dewatering tubes etc) but not including the waste container itself or any added inactive capping material.

waste package

The product of conditioning that includes the wasteform and any container(s) and internal barriers (e.g. absorbing materials and liner), as prepared in accordance with requirements for handling, transport, storage and/or disposal.

waste packager

An organisation responsible for the packaging of radioactive waste in a form suitable for transport and disposal.

Certificate No 4002929

Certificate No 4002929

Nuclear Decommissioning AuthorityRadioactive Waste Management DirectorateBuilding 587Curie AvenueHarwell OxfordDidcotOxfordshire OX11 0RH

+44 (0)1925 802820

+44 (0)1925 802932

www.nda.gov.uk

© Nuclear Decommissioning Authority 2012

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